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About the 1996 TMS Annual Meeting: Tuesday Morning Sessions (February 6)

February 4-8 · 1996 TMS ANNUAL MEETING ·  Anaheim, California


Sponsored by: Jt. SMD/MSD Nuclear Materials Committee

Program Organizers: L. K. Mansur, Metals and Ceramics Division, Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831-6376; C. L. Snead, Jr., Applied Technologies Division, Brookhaven National Laboratory, PO Box 5000, Upton, NY 11973-5000

Tuesday, AM Room: Grand H

February 6, 1996 Location: Anaheim Marriott Hotel

Session Chairperson: L. K. Mansur, Metals and Ceramics Division, Oak Ridge National Laboratory, PO Box 2008, Oak Ridge, TN 37831-6376

8:30 am Invited


Radiation-induced segregation (RIS) occurs in austenitic stainless alloys by enriching Ni and Si and by depleting Cr and Fe at grain boundaries. RIS at grain boundaries is implicated as a cause for irradiation-assisted stress corrosion cracking in fission and fusion reactor technologies. Measured segregation of Fe, Cr and Ni, is consistent with the inverse-Kirkendall vacancy mechanism for a wide range of irradiation conditions. Enrichment of minor elements, Si and P, at grain boundaries is consistent with the interstitial RIS mechanism. However, the kinetics and magnitude of measured Si and P RIS are found to be much less than calculated using the interstitial RIS assumption. In the present paper, an integrated understanding of both major and minor elemental RIS in austenitic stainless alloys is developed. Unique, minor element interaction with defects are identified to rationalize the observed lack of Si and P RIS at grain boundaries. This work was supported by the Materials Sciences Branch, BES, U.S. Department of Energy, under Contract DE-AC06-76RLO 1830.

9:00 am

MODELING OF RADIATION-INDUCED SEGREGATION IN AUSTENITIC Fe-Ni-Cr ALLOYS USING LATTICE RATE THEORY: T. R. Allen, G. S. Was, The University of Michigan, Department of Nuclear Engineering, Ann Arbor, MI 48109

A significant set of radiation-induced segregation (RIS) measurements from proton irradiated and Ni ion irradiated austenitic iron-base and nickel-base alloys now exists. The compositions studied span the iron-rich (304 stainless steel) to nickel-rich (Inconel) regions of the Fe-Cr-Ni phase diagram. Modeling using the Perks vacancy-driven continuum RIS model has shown that to accurately predict segregation trends, alloy specific diffusion parameters must be used. This requirement for alloy-specific diffusion parameters limits the ability of the Perks model to predict segregation in new alloy systems. Additionally, some features of the segregation measurements cannot be explained using simple Inverse Kirkendall behavior. In order to improve on the ability to predict RIS in new alloy systems, a new model for calculating RIS is presented, using lattice rate theory with diffusion parameters based on local atomic configuration. The advantages to a lattice rate model calculation are that migration energies for multiple alloy systems are not needed and that the effects of local order can easily be incorporated into the calculations.

9:20 am

NEUTRON IRRADIATION AND INTERGRANULAR FRACTURE IN V-20 wt.% Ti ALLOYS UNDOPED AND DOPED WITH P OR S: D. Y. Lvu, Chonju Technical College, Korea; T. E. Bloomer, O. Unal, J. Kameda, Ames Laboratory, Iowa State University, Ames, IA 50011

The effect of neutron irradiation (1024 n/m2 at 400deg.C) on the deformation and fracture behavior in V-20 wt.% Ti alloys undoped and doped with P or S has been studied by means of a small punch (SP) testing method in conjunction with scanning Auger microprobe (SAM) and transmission electron microscopy (TEM) analyses. SP tests on unirradiated specimens indicated that the addition of S or P slightly increased the ductile-brittle transition temperature (DBTT). Impurity doped alloys showed a little stronger hardening during the irradiation. The neutron irradiation led to a large increase in the DBTT of a undoped alloy while it did not affect the DBTT in the impurity doped alloys. All the unirradiated and irradiated alloys exhibited smooth or rough intergranular fracture surfaces at the lower shelf regime. The formation of sulfides or phosphides, and segregated S or P at grain boundaries were examined by SAM and TEM. The irradiation effect on the intergranular fracture is discussed in light of defect trapping at incoherent precipitate interfaces in the grain interior and the migration of defect-solute complexes to grain boundaries. The beneficial/detrimental effects of impurities are also discussed in developing fusion reactor materials.

9:40 am

EFFECTS OF RADIATION AND STRAIN ON INTERGRANULAR FRACTURE OF AUSTENITIC STAINLESS STEELS: E. P. Simonen, S. M. Bruemmer, Pacific Northwest Laboratory, P.O. Box 999 MS P8-15, Richland, WA 99352

Intergranular (IG) fracture of irradiated austenitic stainless steels can be promoted by either slow straining at temperatures above 400deg.C in inert environments or by slow straining at lower temperatures in water environments. The synergistic roles of strain and environment on IG fracture are evaluated in the present paper using an Ashby map approach for deformation and fracture kinetics. During post-irradiation straining, radiation retards kinetics for transgranular (TG) processes and provides an opportunity for dominance of IG processes. During in situ straining, radiation creep accelerates transgranular deformation. However, flow of radiation-produced defects to grain boundaries is expected to accelerate IG sliding and promote IG fracture. Radiation effects on IG and TG deformation and fracture are calculated in the present analysis to establish criteria for IG fracture relevant to observed component cracking for both boiling water reactors and pressurized water reactors. This work was supported by the Materials Sciences Branch, Department of Energy, under Contract DE-AC06-76RLO-1830.

10:00 am BREAK

10:20 am Invited


Accelerator-driven neutron technologies include facilities for neutron scattering research, accelerator transmutation of waste, accelerator production of tritium (APT), and accelerator-based conversion of Pu. In these systems, high-energy protons (E > -100 MeV) strike a heavy metal target, producing spallation neutrons with energies extending up to the incident proton energy. Radiation damage (displacements, He, and transmutation products) due to the incident protons and the spallation neutrons have been calculated using the LAHET Code System. For W, a frequently used target material, the displacement cross section is calculated to reach 7300 barns at 800 MeV for both protons and neutrons. For a projected 800 MeV 1 MW neutron source, a peak displacement rate of 93 dpa/yr is calculated in W due to the incident protons and 8 dpa/yr due to spallation neutrons, with 50% of the displacements due to neutrons of E > 20 MeV. The limited experimental information on spallation-radiation-damaged materials will be presented. Comparison with reactor irradiated materials will be discussed.

10:50 am

MODELING FRACTURE TOUGHNESS INDEXED TEMPERATURE SHIFTS IN LOW ACTIVATION MARTENSITIC STAINLESS STEELS: G. R. Odette, Departments of Mechanical Engineering and Materials, University of California, Santa Barbara, CA 93106

Low activation martensitic steels are attractive candidates for fusion applications, but may experience significant embrittlement as a consequence of hardening by fine scale precipitate and defect cluster features that form under irradiation. Fracture toughness (K) as a function of test temperature (T) curves reflect the changing nature of the fracture event with increasing T. At very low T (lower shelf and knee) stressed volume controlled cleavage fracture is initiated in the local plastic zone in front of the crack tip while the specimen is grossly elastic. At higher T (lower transition regime) cleavage occurs following general yielding of the specimen; at still higher T limited stable crack growth takes place prior to cleavage initiation (upper transition regime) by a mixture of stress controlled cleavage and strain controlled microvoid coalescence; and, finally, at high T, fracture takes place by a ductile tearing/microvoid coalescence mechanism. Detailed micromechanical models are needed to relate the crack tip stress strain fields, mediated by the constitutive law, to the local stress-strain-volume conditions leading to fracture. However, for the cleavage regime this complexity might be short circuited by using an empirically based hypotheses that the local conditions for fracture are independent of irradiation and test temperature and that there is a one-to-one correlation between strain hardening and yield stress. In this case cleavage initiation in a given material is simply controlled by the yield stress: thus Kirr (Tirr) = Kunirr (Tunirr) when [[sigma]]yirr=[[sigma]]yunirr where [[Delta]]TK = Tirr - Tunirr (Tirr > Tunirr) represents the temperature shift at some index of toughness (e.g., K = 100 MPam) The micromechanical basis for this hypothesis is described and it is shown that predictions of the simple model are in agreement with limited data from the literature. The results also suggest that T decreases rapidly with Tunirr and increases non-linearly with y. Various limitations of the model are examined. Supported by the Office of Fusion Energy, U. S. Department of Energy Grant DE-FG03-94ER54275.

11:10 am

REFUTATION OF RADIATION SOFTENING IN PRECIPITATION-HARDENED ALUMINUM ALLOYS: K. Farrell, S. T. Mahmood , Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831-6376 (Present address: General Electric Vallecitos Nuclear Center, Pleasanton, CA 94566)

The 6000-series aluminum alloys contain Mg and Si and are hardened by a fine precipitate of Mg2Si, produced by solution treatment, quenching and tempering. Normally, the precipitation-hardened material is stable under displacive irradiation. However, there are two claims of radiation-induced softening, one during neutron irradiation, the other during high energy proton bombardment, in which tensile strengths are reduced to the fully annealed levels and the Mg2Si phase disappears. This paper describes new neutron and proton irradiations that disprove the claims and indicate that the reported softening is caused by thermal overaging due to inadequate control of irradiation heating. Research sponsored by the Division of Materials Sciences, U.S. Department of Energy under contract DE-AC05-840R21400 with Lockheed Martin Energy Systems. One of the authors (S. T. Mahmood) was supported by an appointment to the Oak Ridge National Laboratory Postdoctoral Research Program administered by the Oak Ridge Institute for Science and Education.

11:30 am

FORMATION OF Kr BUBBLES IN ZIRCONIUM ALLOYS: L. Pagano, A. T. Motta, Department of Nuclear Engineering, The Pennsylvania State University, University Park, PA 16802

Zirconium alloys are remarkably resistant to void swelling and bubble formation, but bubbles can be formed during irradiation when stabilized by insoluble gases. Several zirconium alloys, including Zircaloys and pure Zr, were irradiated with 100 keV Kr ions in the HVEM/Tandem facility at Argonne National Laboratory at temperatures ranging from room temperature to 800deg.C. The objective was to conduct a systematic study of the conditions for bubble development under irradiation, and study the influence of alloy microstructure and composition. Results indicate that a gas-driven regime, (where a fine dispersion of 30-60 bubbles is seen), exists from room temperature up to 500deg.C. At higher temperatures (above 700deg.C) larger faceted bubbles (about 500deg.) were observed, as a result of bubble mobility and coalescence. The results are discussed in terms of existing bubble growth models.

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